In the operation of liquid sodium or sodium-potassium metal cooled nuclear reactors for power generation, it may be necessary to shut down the fission reaction of the fuel to deal with emergencies or carry out routine maintenance services. Reactor shut down is attained by inserting neutron absorbing control rods into the core of fissionable fuel to deprive the fuel of the needed fission producing neutrons. However decay of the fuel in the shut down reactor continues to produce heat in significant amounts which must be dissipated from the reactor unit.
The heat capacity of the liquid metal coolant and adjacent structure aid in dissipating the residual heat. However, the structural materials of the nuclear reactor may not be capable of safely withstanding prolonged high temperatures. For example the concrete of the walls of the typical housing silo may splay and crack when subjected to high temperatures. Accordingly, auxiliary cooling systems are commonly utilized to safely remove heat from the nuclear reactor structure during shut down.
Conventional nuclear reactors have utilized a variety of elaborate energy driven cooling systems to dissipate heat from the reactor. In many of the situations warranting a shutdown, the energy supply to the cooling systems make the cooling systems themselves subject to failure. For example, pumps and ventilation systems to cool the core may fail. Furthermore, if operator intervention is necessary, there are foreseeable scenarios in which the operator would be unable to provide the appropriate action. The most reliable and desirable cooling system would be a completely passive system which could continuously remove the residual heat generated after shutdown regardless of conditions.
Liquid metal cooled reactors such as the modular type disclosed in U.S. Pat. No. 4,508,677, utilizing sodium or sodium-potassium as the coolant provides numerous advantages. Water cooled reactors operate at or near the boiling point of water. Any significant rise in temperature results in the generation of steam and increased pressure. By contrast, sodium or sodium-potassium has an extremely high boiling point, in the range of 1800 degrees Fahrenheit at one atmosphere pressure. The normal operating temperature of the reactor is in the range of about 900 degrees Fahrenheit. Because of the high boiling point of the liquid metal, the pressure problems associated with water cooled reactors and the steam generated thereby are eliminated. The heat capacity of the liquid metal permits the sodium or sodium-potassium to be heated several hundred degrees Fahrenheit without danger of materials failure in the reactor.
The reactor vessels for pool-type liquid-metal cooled reactors are essentially open top cylindrical tanks without any perforations to interrupt the integrity of the vessel walls. Sealing of side and bottom walls is essential to prevent the leakage of liquid metal from the primary vessel. The vessel surfaces must also be accessible for the rigorous inspections required by safety considerations.
In the typical sodium cooled reactor, two levels of heat conveying sodium loops or cooling circuits are used. Usually, a single primary loop and two or more secondary loops are used. The primary heat transferring loop contains very radioactive sodium which is heated by the fuel rods. The primary loop passes through heat exchangers to exchange the heat with one of the non-radioactive secondary sodium loops. In general, sodium cooled reactors ar designed to incorporate redundant secondary heat transferring loops in the event of failure of one loop.
Upon shutdown of the reactor by fully inserting the control rods, residual heat continues to be produced and dissipated according to the heat capacity of the plant. Assuming that the reactor has been at full power for a long period of time, during the first hour following shutdown, an average of about 2% of full power continues to be generated. The residual heat produced continues to decay with time.
Exaggerated conservative safety concerns for dealing with postulated worst possible scenario accident conditions have raised questions as to means for coping with events such as the coincidental failure of both the reactor vessel and the containment or guard vessel, whereupon liquid coolant loss due to the resulting leakage could significantly lower the coolant level within the reactor vessel. Reduced reactor coolant levels can significantly impede or interrupt the normal coolant circulation flow through a coolant loop or circuit, whereby heat is transported away from the fuel core during routine operation. This impediment or termination due to reduced coolant level also applies to designed passive cooling systems employing inherent processes comprising the natural convection of fluids, conduction, radiation and convection, as a means of removing heat through its transfer by such means. Other such improbable extreme events possible affecting coolant levels include a hypothetical core disassembly accident that damages the fuel core and results in expulsion of coolant such as sodium up into the head access area of the reactor structure, or a maintenance accident involving a break in the reactor closure head.
This invention comprises improvements in safety systems for coping with shutdown decay heat from a liquid metal cooled nuclear reactor such as the unit disclosed and claimed in U.S. Letters Pat. No. 4,678,626, issued Dec. 2, 1985.
The disclosed contents of the above noted U.S. Letters Pat. Nos. 4,508,677 and 4,678,626, comprising related background art, are incorporated herein by reference.